General Corrosion Behavior of Alloy 800 Steam Generator Tube under Different Water Chemistry Conditions
摘 要
通过高温高压循环水均匀腐蚀试验及电化学测试方法,研究了800合金传热管在300 ℃、10 MPa、去离子水、溶氧(DO=2 mg/L)以及溶氢(DH=2.7 mg/L)条件下的均匀腐蚀速率及表面氧化膜特征。结果表明,800合金在DO及DH这两种条件下的均匀腐蚀速率均很低,分别为5.6 μm/a和1.2 μm/a。在这两种环境中,试样氧化膜均为内外双层结构,内层连续而致密,外层主要由颗粒状氧化物组成。800合金在正常运行环境(DH工况)中,具有较稳定的电化学特征和明显的钝化区;而在含氧环境中(DO=2 mg/L),随着时间的延长,在96 h后,开路电位会发生明显的升高,开路电位由-0.39 V(vs.SHE)逐渐升高至0.42 V(vs.SHE),呈现出无明显的钝化区,但自腐蚀电流密度未有明显提高,且阻抗值较大,说明合金表面仍具有较稳定的氧化膜结构。
Abstract
Through high temperature and high pressure circulating water test and electrochemical test, the uniform corrosion rate and surface oxide film characteristics of alloy 800 tube were studied under the conditions of 300 ℃, 10 MPa, deionized water, 2 mg/L DO (dissolved oxygen) and 2.7 mg/L DH (dissolved hydrogen). The results showed that the uniform corrosion rates of alloy 800 under DO and DH conditions were very low, which were 5.6 μm/a and 1.2 μm/a respectively. The oxide films under these two conditions were both composed of two layers. The inner layer of the oxide film was continuous and dense, while the outer layer was mainly composed of granular oxide. In the normal operation environment (DH working condition), alloy 800 presented more stable electrochemical characteristics and obvious passivation zone; while in the oxygen containing environment (DO=2 mg/L), the open circuit potential increased significantly after 96 h, increased from -0.39 V (vs. SHE) to 0.42 V (vs. SHE). No obvious passivation was observed, but the self-corrosion current density was not increased and the impedance value was much higher, indicating that the alloy surface still had relatively stable oxide film structure.
中图分类号 TG174 DOI 10.11973/fsyfh-202009004
所属栏目 核电设备的腐蚀与防护
基金项目 大型先进压水堆核电重大专项(2018ZX06001001)
收稿日期 2020/5/15
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引用该论文: LIU Xiaoqiang,LIU Ji,SUN Wei,ZHANG Zhiming,MENG Fanjiang,WANG Jianqiu,SHI Xiuqiang. General Corrosion Behavior of Alloy 800 Steam Generator Tube under Different Water Chemistry Conditions[J]. Corrosion & Protection, 2020, 41(9): 20
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参考文献
【1】STAEHLE R W,GORMAN J A. Quantitative assessment of submodes of stress corrosion cracking on the secondary side of steam generator tubing in pressurized water reactors:part 1[J]. Corrosion,2003,59(11):931-994.
【2】PANDEY M D,DATLA S,TAPPING R L,et al. The estimation of lifetime distribution of Alloy 800 steam generator tubing[J]. Nuclear Engineering and Design,2009,239(10):1862-1869.
【3】BERGANT M A,YAWNY A A,PEREZ IPIÑA J E. A comparison of failure assessment diagram options for Inconel 690 and Incoloy 800 nuclear steam generators tubes[J]. Annals of Nuclear Energy,2020,140:107310.
【4】LI X H,WANG J Q,HAN E H,et al. Corrosion behavior for Alloy 690 and Alloy 800 tubes in simulated primary water[J]. Corrosion Science,2013,67:169-178.
【5】CHEUNG C,ERB U,PALUMBO G. Application of grain boundary engineering concepts to alleviate intergranular cracking in alloys 600 and 690[J]. Materials Science and Engineering:A,1994,185(1/2):39-43.
【6】Steam Generator Management Program:Alloy 800 steam generator tubing experience[R]. Palo Alto Electric Power Research Institute,2012.
【7】ARIOKA K,YAMADA T,MIYAMOTO T,et al. Intergranular stress corrosion cracking growth behavior of Ni-Cr-Fe alloys in pressurized water reactor primary water[J]. Corrosion,2014,70(7):695-707.
【8】GORMAN J A,MORONEY V D,WHITE G. Alloy 800 steam generator tube performance[C]//Proceedings of the 6th International Steam Generator Conference,Toronto:[s.n.],2009.
【9】KOMAR M L,GOSZCZYNSKI G. Stress corrosion cracking of alloy 800 in secondary side crevice environment[M]//The Minerals,Metals & Materials Series.Cham:Springer International Publishing,2017:2409-2419.
【10】LUCAN D,FULGER M,JINESCU G. Corrosion process of incoloy-800 in high pressure and temperature aqueous environment[J]. Revista De Chimie,2008,59(9):1026-1029.
【11】WANG J Q,HUANG F,KE W. Corrosion behaviors of Inconel 690TT and Incoloy 800MA steam generator tubes in high temperature high pressure water[J]. Acta Metall Sin,2016,52(10):1333-1344.
【12】ZHANG Z M,WANG J Q,HAN E H,et al. Analysis of surface oxide film formed on eletropolished alloy 690TT in high temperature and high pressure water with sequentially dissolved hydrogen and oxygen[J]. Acta Metall Sin,2015,51(1):85-92.
【13】LI X H,WANG J Q,HAN E H,et al. Corrosion behavior of nuclear grade alloys 690 and 800 in simulated high temperature and high pressure primary water of pressurized water reactor[J]. Acta Metall Sin,2012,48(8):941.
【14】WANG J Q,LI X H,HUANG F,et al. Comparison of corrosion resistance of UNS N06690TT and UNS N08800SN in simulated primary water with various concentrations of Dissolved oxygen[J]. Corrosion,2014,70(6):598-614.
【15】MACÁK J,SAJDL P,KUCERA P,et al. In situ electrochemical impedance and noise measurements of corroding stainless steel in high temperature water[J]. Electrochimica Acta,2006,51(17):3566-3577.
【16】BOSCH R W,WÉBER M,VANKEERBERGHEN M. In-pile electrochemical measurements on AISI 304 and AISI 306 in PWR conditions-Experimental results[J]. Journal of Nuclear Materials,2007,360(3):304-314.
【2】PANDEY M D,DATLA S,TAPPING R L,et al. The estimation of lifetime distribution of Alloy 800 steam generator tubing[J]. Nuclear Engineering and Design,2009,239(10):1862-1869.
【3】BERGANT M A,YAWNY A A,PEREZ IPIÑA J E. A comparison of failure assessment diagram options for Inconel 690 and Incoloy 800 nuclear steam generators tubes[J]. Annals of Nuclear Energy,2020,140:107310.
【4】LI X H,WANG J Q,HAN E H,et al. Corrosion behavior for Alloy 690 and Alloy 800 tubes in simulated primary water[J]. Corrosion Science,2013,67:169-178.
【5】CHEUNG C,ERB U,PALUMBO G. Application of grain boundary engineering concepts to alleviate intergranular cracking in alloys 600 and 690[J]. Materials Science and Engineering:A,1994,185(1/2):39-43.
【6】Steam Generator Management Program:Alloy 800 steam generator tubing experience[R]. Palo Alto Electric Power Research Institute,2012.
【7】ARIOKA K,YAMADA T,MIYAMOTO T,et al. Intergranular stress corrosion cracking growth behavior of Ni-Cr-Fe alloys in pressurized water reactor primary water[J]. Corrosion,2014,70(7):695-707.
【8】GORMAN J A,MORONEY V D,WHITE G. Alloy 800 steam generator tube performance[C]//Proceedings of the 6th International Steam Generator Conference,Toronto:[s.n.],2009.
【9】KOMAR M L,GOSZCZYNSKI G. Stress corrosion cracking of alloy 800 in secondary side crevice environment[M]//The Minerals,Metals & Materials Series.Cham:Springer International Publishing,2017:2409-2419.
【10】LUCAN D,FULGER M,JINESCU G. Corrosion process of incoloy-800 in high pressure and temperature aqueous environment[J]. Revista De Chimie,2008,59(9):1026-1029.
【11】WANG J Q,HUANG F,KE W. Corrosion behaviors of Inconel 690TT and Incoloy 800MA steam generator tubes in high temperature high pressure water[J]. Acta Metall Sin,2016,52(10):1333-1344.
【12】ZHANG Z M,WANG J Q,HAN E H,et al. Analysis of surface oxide film formed on eletropolished alloy 690TT in high temperature and high pressure water with sequentially dissolved hydrogen and oxygen[J]. Acta Metall Sin,2015,51(1):85-92.
【13】LI X H,WANG J Q,HAN E H,et al. Corrosion behavior of nuclear grade alloys 690 and 800 in simulated high temperature and high pressure primary water of pressurized water reactor[J]. Acta Metall Sin,2012,48(8):941.
【14】WANG J Q,LI X H,HUANG F,et al. Comparison of corrosion resistance of UNS N06690TT and UNS N08800SN in simulated primary water with various concentrations of Dissolved oxygen[J]. Corrosion,2014,70(6):598-614.
【15】MACÁK J,SAJDL P,KUCERA P,et al. In situ electrochemical impedance and noise measurements of corroding stainless steel in high temperature water[J]. Electrochimica Acta,2006,51(17):3566-3577.
【16】BOSCH R W,WÉBER M,VANKEERBERGHEN M. In-pile electrochemical measurements on AISI 304 and AISI 306 in PWR conditions-Experimental results[J]. Journal of Nuclear Materials,2007,360(3):304-314.
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